Estimation of magnetic error field with alleviating fabrication tolerance of large superconducting magnets on JA DEMO reactor

Generally, DEMO requires larger toroidal field (TF) coils than ITER, resulting in one of the major difficulties, the tolerance in TF coil fabrication. This paper presents the possible solutions based on the design study on Japan’s DEMO (JA DEMO). It was confirmed that, in the case of adopting a mitigated tolerance by a factor of 2.5–5 compared with that of ITER, the resulting error field of TF coils is correctable to an acceptable level in terms of locked mode avoidance. In addition, the design of the error field correction coil (EFCC) on JA DEMO was investigated

Design strategy and recent design activity on Japan\u27s DEMO

The Joint Special Design Team for Fusion DEMO was organized in 2015 to enhance Japan’s DEMO design activity and coordinate relevant research and development (R&D) toward DEMO. This paper presents the fundamental concept of DEMO and its key components with main arguments on DEMO design strategy. Superconducting magnet technology on toroidal field coils is based on the ITER scheme where a cable-in-conduit Nb3Sn conductor is inserted in the groove of a radial plate. Development of cryogenic steel with higher strength is a major challenge on the magnet. Divertor study has led to a baseline concept based on water-cooled single-null divertor assuming plasma detachment. Regarding breeding blanket, fundamental design study has been continued with focuses on tritium self-sufficiency, pressure tightness in case of in-box LOCA (loss of coolant accident) and material compatibility. An important finding on tritium permeation to the cooling water is also reported, indicating that the permeation to the cooling water is manageable with existing technology